Experiments on the interactions of molten ZrO2 with water using TROI facility

Korea Atomic Energy Research Institute (KAERI) launched an intermediate scale steam explosion experiment named ‘Test for Real cOrium Interaction with water (TROI)’ using reactor material. The objective of the program is to investigate whether the corium would lead to energetic steam explosion when interacted with cold water at a low pressure. The melt/water interaction is made in a multi-dimensional test section located in a pressure vessel. The inductive skull melting, which is basically a direct inductive heating of an electrically conducting melt, is implemented for the melting and delivery of corium. In the first series of tests using several kg of ZrO₂ where the melt/water interaction is made in a heated water pool at 30–95 °C, either a quenching or a spontaneous steam explosion was observed. The spontaneous explosion observed in the present ZrO₂ melt/water experiments clearly indicates that the physical properties of the UO₂/ZrO₂ mixture have a strong effect on the energetics of steam explosion.     

The prediction of 2D thermal detonations and resulting damage potential

The main purpose of this paper is to introduce a new concept for the processes responsible for the escalation and propagation of steam explosions. The concept recognizes that initially only a small quantity of coolant around each coarsely premixed melt mass “sees” the fragmenting debris coming off it, hence it is called the concept of “microinteractions”. We also derive the analytical basis for it, define the nature of the requisite constitutive laws and related experimental data, and demonstrate that this concept is essential for the prediction of steam explosion energetics in large-scale premixtures in 2D geometries. We also provide the first numerical illustrations of this concept, implemented in the computer code .m. Further, we provide the first numerical results of steam explosions in large water pools, i.e. ex-vessel explosions. These results reveal two important mechanisms for explosion “venting” and thus for reducing the dynamic loads on adjacent structures. We conclude that, taken together, the “microinteractions” and “venting” make realistic predictions of steam explosion loads feasible and within reach in the near future.     

Results of recent KROTOS FCI tests: alumina versus corium melts

Recent results from KROTOS fuel-coolant interaction experiments are discussed. Five tests with alumina were performed under highly subcooled conditions, all of these tests resulted in spontaneous steam explosions. Additionally, four tests were performed at low subcooling to confirm, on one hand, the suppression of spontaneous steam explosions under such conditions and, on the other hand, that such a system is still triggerable using an external initiator. The other test parameters in these alumina tests included the melt superheat and the initial pressure. All the tests in the investigated superheat range (150–750 K) produced a steam explosion and no evidence of the explosion suppression by the elevated initial pressure (in the limited range of 0.1–0.375 MPa) was observed in the alumina tests. The corium test series include a test with 3 kg of melt under both subcooled and near saturated conditions at ambient pressure. Two additional tests were performed with subcooled water; one test was performed at an elevated pressure of 0.2 MPa with 2.4 kg of melt and another test with 5.1 kg of melt at ambient pressure. None of these tests with corium produced a propagating energetic steam explosion. However, propagating low energy (about twice the energy of the trigger pulse) events were observed. All corium tests produced significantly higher water level swells during the mixing phase than the corresponding alumina tests. Present experimental evidence suggests that the water depletion in the mixing zone suppresses energetic steam explosions with corium melts at ambient pressure and in the present pour geometry. Processes that could produce such a difference in void generation are discussed.     

Effect of melt water interaction configuration on the process of steam explosion

Steam explosion experiments are performed at various modes of melt water interaction configuration using prototypic corium melt. The tests are performed to simulate both melt water interaction in a partially flooded cavity and melt water interaction in a cavity with submerged reactor. The tests are performed using zirconia and corium melts. The behavior of melt jet fragmentation during the flight in the air and fragmentation and mixing of melt jet in water is investigated by a high-speed video visualization and by comparison of debris size distribution and morphology of debris. Strength of steam explosion is estimated by measuring dynamic pressure and dynamic force.     

FCI experiments in the corium/water system

The KROTOS fuel coolant interaction (FCI) tests are aimed at providing benchmark data to examine the effect of fuel/coolant initial conditions and mixing on explosion energetics. Experiments, fundamental in nature, are performed in well-controlled geometries and are complementary to the FARO large scale tests. Recently, a test series was performed using 3 kg of prototypical corium (80 w/o UO₂, 20 w/o ZrO₂) which was poured into a water column of ≤1.25 m in height (95 and 200 mm in diameter) under 0.1 MPa ambient pressure. Four tests were performed in the test section of 95 mm in diameter (ID) with different subcooling levels (10–80 K) and with and without an external trigger. Additionally, one test has been performed with a test section of 200 mm in diameter (ID) and with an external trigger. No spontaneous or triggered energetic FCIs (steam explosions) were observed in these corium tests. This is in sharp contrast with the steam explosions observed in the previously reported alumina (Al₂O₃) test series which had the same initial conditions of ambient pressure and subcooling. The post-test analysis of the corium experiments indicated that strong vaporisation at the melt/water contact led to a partial expulsion of the melt from the test section into the pressure vessel. In order to avoid this and to obtain a good penetration and premixing of the corium melt, an additional test was performed with a larger diameter test section. In all the corium tests an efficient quenching process (0.8–1.0 MW kg-melt⁻¹) with total fuel fragmentation (mass mean diameter 1.4–2.5 mm) was observed. Results from alumina tests under the same initial conditions are also given to highlight the differences in behaviour between corium and alumina melts during the melt/water mixing.     

Inherently safe containments for nuclear power plants

Nuclear energy cannot be avoided in the near future. To regain public acceptance the safety of nuclear power plants has to be increased. Consequently, feasibility studies have been carried out for a containment proposal for future pressurized water reactors which will keep people unharmed even in the case of severe nuclear accidents under the assumption “all that can go wrong, will go wrong”. The main features of the design concept are a core melt cooling and retention device, a passively acting cooling system to remove the decay heat and a double-wall containment which is able to withstand high static and dynamic internal pressures due to hydrogen detonation. Internal structures are designed to resist extreme loadings resulting from various accident scenarios including in-vessel steam explosion and vessel failure under high system pressure.     

The effect of corium composition and interaction vessel geometry on the prototypic steam explosion

This paper discusses the results of steam explosion experiments using molten material consisting of UO₂ and ZrO₂ mixture, which is called corium, to simulate a prototypic steam explosion in a nuclear reactor during a postulated severe accident. About 5–10 kg of molten material with enough superheat was poured into a pool of water in a test section at room temperature to simulate ex-vessel steam explosion in the reactor situation. Most of the experiments were externally triggered. The purpose of the experiments was to investigate the effect of material composition and average void fraction on the strength of a prototypic steam explosion, which were highlighted as major unresolved issues.The experiments were performed using two kinds of mixtures, one, corium A, at 70:30 weight percent composition of UO₂ and ZrO₂, close to eutectic composition, and the other, corium B, at 80:20 weight percent. Also, two kinds of cylindrical test sections having a different diameter were used. It turned out that corium A was likely to produce an energetic steam explosion, while corium B seldom led to an energetic steam explosion. The existence of mush phase for the non-eutectic mixture is suggested to be the reason for the difference. Comparative cross sectional views of the corium particles by scanning electron microscope supported the proposed argument. The tests performed with a narrow test section seldom led to an energetic steam explosion for both materials. An increase in average void is suggested to be the reason for the non-explosive behavior, which is consistent with the physical models employed in the current steam explosion computer codes.     

FCI experiments in the aluminum oxide/water system

The KROTOS facility at JRC Ispra was recently used to study experimentally melt-coolant premixing and steam explosion phenomena in Al₂O₃/water mixtures with approximately 1.5 kg melt at 2300–2400 °C. In the five tests performed the main parameter was the water subcooling, 10, 40 and 80 X, respectively. In the nearly saturated system, steam explosions could be externally triggered, which resulted in high (supercritical) explosion pressures in the test tube: KROTOS 26, 28. Without triggering, melt penetration in water and melt agglomeration on the bottom plate of the test tube could be observed, which gave rise to strong steaming during the melt cooling-down process: KROTOS 27. In the two tests KROTOS 29, 30, performed with 80 K subcooled water, self-triggered steam explosions occurred with pressures of more than 100 MPa. Post-test analysis of the debris revealed that 85% of the interacting fuel mass fragmented in particles of sizes smaller than 250 μm. An energy conversion ratio of 1.25% was estimated from vessel pressurization data taking into account the energy content in the fuel mass which fragmented to particle diameters of less than 250 μm. The test section was damaged in the test KROTOS 30.     

Transient deformation of LMFBR cores due to local failure: Experimental and theoretical investigation

This paper describes an effort to predict the mechanical core deformation caused by local failure within an LMFBR core. These activities are intended to cover all the potential core damage possibilities currently under discussion and analysis. In particular it is shown that the reactor can be scrammed in time under pessimistic-realistic pressure transients and that the damage does not exceed tolerable limits.A special gas generator technique to simulate a fuel coolant explosion has been developed at AWRE Foulness. This has been used to perform the explosion tests needed to demonstrate the safety of the SNR 300 core. A molten fuel—coolant interaction (MFCI) experimental facility, and a drop tower to carry out sub-assembly crushing tests are described. Theoretical studies are presented which use mass-spring-dashpot, lumped parameter-hinge or micro-rigid-lumped-mass models. They simulate the crushing and bending of a single sub-assembly interacting with the coolant as well as the behaviour of a multirow “spoke” model.For the core analysis only preliminary computational results are presently available which can be compared with the full scale tests in which the fluid pressure did not exceed a “threshold” of about 100 bar. Parameter studies show the influence of pulse shape, material properties as well as the time integrator.Some of the unanswered question concern the dydrodynamic feedback of the deformations on the pressure distribution in space and time. Also the behaviour of the highly irradiation-embrittled cores is poorly understood today. Finally, an enhanced energy release package to describe the MFCI must still be added to the reactivity calculation module of a future fast reactor dynamics code.     

Estimation of ex-vessel steam explosion pressure loads

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment.In this article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D, which is being developed for the simulation of fuel–coolant interactions. A parametric study was performed varying the location of the melt release (central, right and left side melt pour), the cavity water subcooling, the primary system overpressure at vessel failure and the triggering time for explosion calculations. The main purpose of the study was to establish the influence of the varied parameters on the fuel–coolant interaction behaviour, to determine the most challenging cases and to estimate the expected pressure loadings on the cavity walls. For the most explosive central, right side and left side melt pour scenarios a detailed analysis of the explosion simulation results was performed. The study shows that for some ex-vessel steam explosion scenarios higher pressure loads are predicted than obtained in the OECD programme SERENA phase 1.     

The DEEPSSI project, design, testing and modeling of steam injectors

The DEEPSSI project is a steam injector research programme. Among thermal-hydraulic passive systems, the steam injectors (also called “condensing ejectors” or “steam jet pumps”) are very interesting apparatus with very specific characteristics (high velocity, very low pressure). The envisaged reactor application is the Steam Generator Emergency FeedWater System (EFWS) of Pressurised Water Reactors (PWRs). The heart of this project is the development and the testing of an innovative steam injector design. Three experimental facilities are involved: CLAUDIA in France, IETI in Italy and IMP-PAN in Poland. In these facilities, different design options have been tested and some significant improvements of the initial design have been obtained.In addition to the experimental studies, the development of a steam injector computational model has been undertaken in order to model industrial systems based on steam injectors. The one-dimensional module of the system code CATHARE2 has been chosen to be the basis of this model. The first results obtained have confirmed the capabilities of CATHARE2 to describe the steam injector thermal-hydraulics.     

Heat transfer from a partially uncovered reactor core

This paper documents a model which has been developed for predicting the temperature distribution along a “flow channel” of a pressurized water reactor during simulated, uncovered core conditions. In the model, heat conduction along the fuel element, convection from the surface to the coolant, radiation exchange between the clad surface and steam, and surface exchange between adjacent fuel rods are considered. Variations of the thermophysical properties of the fuel road and of the coolant with temperature are accounted for, but oxidation of Zircaloy is not modeled. Extensive sensitivity studies on the effects of heat generation in the core, steam velocity, pressure level, uncovered core height, presence of hydrogen gas in the coolant, power skew, clad emissivity, and convective heat transfer correlations have been examined. The results show that the importance of radiation in comparison with convection increases with an increase in the fuel rod temperature, pressure, and clad emissivity.     

Dynamic model for control system design and simulation of a low temperature nuclear reactor

A lumped parameter dynamic model for the primary-loop and the U-tube steam generator of a low temperature power reactor is developed based on the fundamental conservation laws of fluid mass, energy and momentum. The dynamic model is formulated by coupling the point kinetics with reactivity feedback and the thermal-hydraulics of the reactor. The developed dynamic model is implemented on a personal computer using MATLAB/SIMULINK. Numerical simulation results for steady-state and transient responses are then presented, which show that the steady-state precision of the newly developed dynamic model is acceptable and the trend of the transient responses is correct. In addition, the “swell and shrink” behavior of the U-tube steam generator is also verified by numerical simulation. This newly established model can be utilized to control system design and simulation for the low temperature power reactor.     

Molecular-sieving effect of zeolite 3A on adsorption of H2, HD and D2

Synthetic zeolite 3A has the molecular-sieving windows of nominal diameter 0.3 nm in its crystal lattice framework, which obstruct the crystalline adsorption of molecules of diameter larger than 0.3 nm, except water, hydrogen and helium. The window's diameter slightly varies with temperature, however, that is endorsed in experimental results that hydrogen cannot be adsorbed at the liquid-nitrogen temperature. Authors measured the range of temperature where zeolite 3A permits hydrogen adsorption, and revealed the temperature difference of several degrees in appearance of molecular sieving for H₂ and D₂. This difference is important because from a H₂–D₂ mixture one isotope could be isolated by adsorption if operated at a temperature regulated between the molecular-sieving appearance temperatures. We have reported large values of D₂/H₂ separation factor obtained from molecular-sieving experiments. In this study, the effect of sieving for the hybrid-atomic isotope HD is examined using a H₂–HD–D₂ mixture. We here report the experimental HD/H₂ separation factor evaluated between the D₂/H₂ factor and unity. This result is significant because where the effective molecular diameter concerning the sieving mechanism is suggested. From this knowledge, the isotopic effect of sieving for HT and DT can be predicted.     

QUEOS, an experimental investigation of the premixing phase with hot spheres

A total of 60 experiments with hot spheres, to investigate the premixing phase of a steam explosion, have been performed in the QUEOS facility at Forschungszentrum Karlsruhe (FZK). Here, the data of seven experiments are presented using three types of spheres at 1800 K and total volumes of ≈2, and 4l, respectively. The sphere jet plunging into the water had a diameter of 10 cm and a length between 60 and 120 cm. The average solid volume fraction was ≈25%. These relatively long jets and high particle volume fractions show a different behavior in the water, compared to the 18 cm wide short pours of a first series of QUEOS experiments. They are also unlike experiments performed in other facilities with much lower particle volume fractions. High speed films were taken, pressures, water temperatures and the steaming rate were measured.     

UPTF experiment refined PWR LOCA thermal-hydraulic scenarios: conclusions from a full-scale experimental program

A total of 34 tests were performed at upper plenum test facility (UPTF, a 1:1 scale test facility) to investigate the thermal-hydraulic phenomena in a pressurized water reactor (PWR) primary system during end-of-blowdown, refill and reflood phases of a loss-of-coolant accident (LOCA). Separate effect tests as well as integral tests were carried out. After the completion of the program a summary of the basic findings from the full-scale tests is given, focusing on thermal-hydraulic issues related to: two-phase flow phenomena at the ECC injection ports for cold or hot leg injection; the ECC delivery into core area via the downcomer or the tie plate; entrainment de-entrainment phenomena during reflood (i.e. the “steam binding” and driving water head reduction problems).     

Steam generator tube vibration tests

The buildup of magnetite in the steam generators of some pressurized water reactors had led operators to propose chemical cleaning to remove this product. In some cases, the volume of magnetite formed by the corrosion of the carbon steel has been sufficient to cause “denting” or reduction of the outer diameter of the tubes where they pass through the support plates. The US Nuclear Regulatory Commission has expressed concern that the resulting increased clearances may allow an increased level of flow-induced vibrations in chemically cleaned steam generators that could lead to high tube wear rates and unacceptable levels of tube failure.A project has been undertaken by Pacific Northwest Laboratory to address the effects of increased tube/tube-support clearances and to provide the NRC with criteria to evaluate licensees' specific proposals for chemical cleaning of steam generators. The first phase of the project consists of flow tests to establish the forcing boundary conditions, using clearances representing various conditions following chemical cleaning. The second phase consists of accelerated wear tests to determine the potential wear rates possible based on the vibrations characterized in the flow tests.     

Significance of fluid-structure interaction phenomena for containment response to ex-vessel steam explosions

When studying the structural response of a containment building to ex-vessel steam explosion loads, a two-step procedure is often used. In the first step of this procedure the structures are treated as rigid and the pressure-time history generated by the explosion, at the rigid wall, is calculated. In the second step the calculated pressure is applied to the structures. The obvious weakness of the two-step procedure is that it does not correspond to the real dynamic behaviour of the fluid-structure system. The purpose of this paper is to identify and evaluate the relevant fluid-structure interaction phenomena. This is achieved through direct treatment of the explosion process and the structural response. The predictions of a direct and two-step treatment are compared for a BWR Mark II containment design, consisting of two concentric walls interacting with water masses in the central and annular pools. It is shown that the two-step approach leads to unrealistic energy transfer in the containment system studied and to significant overestimation of the deflection of the containment wall. As regards the pedestal wall, the direct method analysis shows that the flexibility of this wall affects the pressure-time history considerably. Three load types have been identified for this wall namely shock load, water blow as a result of water cavitation, and hydrodynamic load. Reloading impulse due to cavitation phenomena plays an important role as it amounts to ≈40% of the total impulse load. Investigation of the generality of the cavitation phenomena in the context of ex-vessel steam explosion loads was outside the scope of this work.     

Optical properties of UO2 and PuO2

We perform first-principles calculations of electronic structure and optical properties for UO₂ and PuO₂ based on the density functional theory using the generalized gradient approximation (GGA) + U scheme. The main features in orbital-resolved partial density of states for occupied f and p orbitals, unoccupied d orbitals, and related gaps are well reproduced compared to experimental observations. Based on the satisfactory ground-state electronic structure calculations, the dynamical dielectric function and related optical spectra, i.e., the reflectivity, adsorption coefficient, energy-loss, and refractive index spectrum, are obtained. These results are consistent with the available experiments.